ML20202E320
| ML20202E320 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/06/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20202E294 | List: |
| References | |
| 50-313-97-21, NUDOCS 9802180119 | |
| Download: ML20202E320 (3) | |
Text
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ENCLOSURE.1.
NOTICE OF VIOLATION Entergy Operations, Inc.
Docket No.:
50-313 Arkansas Nuclear One License No.: DPR-51 During an NRC inspection conducted on October 27-31 and November 10-14,1997,- two volations of NRC requirements were identified, in accordaace with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:
A.
'O CFR Part 50, Appendix B, Criterion V, requires that activities affecting quality be prescribed by and performed in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances.
Contrary to the above, as of November 23,1991, Procedure 1010.002, " Transient History / Transient Cycle Logging," was inappropriate to the circumstances in that a vendor-recommended maximum emergency feedwater flow limit of 1500 gpm to a single -
steam generator was not incorporated into the procedure.
This is a Severity Level IV violation (Supplement 1) (50-3*.3/9721-01).
B. -
10 CFR Part 50, Appendix B, Criterion 111, " Design Control," states that measures shall I
be established to ensure that applicable requirements and the design basis are correctly
- translated into specifications, drawings, procedures, and instructions and that the design control measures shall provide for verifying and checking the adequacy of the design, such as by the performance of design reviews.
. Contrary to the above, in the following seven examples, the adequacy of the design, as defined in engineering calculations, was not verified and/or the design basis of the plant was not correctly translated into the installed configuration:
1.
The piping configuration design inputs for Calculation 82-D-2086-02, Revisinn 3, dated September 22,1986, were not properly verified in that the piping -
configuration used in the calculation of pressure drops was not consistent with the piping isometric drawings.
2.
Calculation 88-E-0086-01, "lE Bulletin 88-04 Review for the P7A and P7B Minimum Flow Evaluation," Revision 0, dated June 2,1989, performed to ensure appropriate minimum flow for the emergency feedwater pumps, was not properly verified for adequacy of the design in that the effects of minimum recirculation during operation of both emergency feedwater pumps in parallel was not documented.
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2-3.
Calculation 9? ' 0077-04, " Unit i EFW System Pump Performance Requirements,' Revision 0, dated November 30,1992, was not properly verified for adequacy of the design in that it used an incorrect level in the condensate storage tank to determine the pump developed head in the case of an emergency feedwater actuation due to a loss of offsite power following a tornado.
4.
Calculation 92 E-0021-01, " Emergency Duty Cycle and Battery Sizing Calculation," Revision 4, dated September 16,1996, was not properly verified for adequacy of the design in that it was not revised to reflect plant modifications that resulted in changes in the continuous and locked rotor current of Emergency Feedwater Valves CV-2620 and CV-2627 and a plant modification that resulted in an additionalload due to the reactor building spray pump closo and trip currents.
5.
The design control measures did not ensure that Calculation 80-D-1083A-02, "EFIC DC Valve Torque Calculation Under Reduced Voltage Condition,"
Revision 1, dated April 2,1986, was adequately checked in that the calculation was superseded but not designated as such.
6.
Since initial licensing, measures failed to ensure that seismic design requirements, as de.,ineated in seismic drawing details, were correctly translated into steam generator instrument tubing installations and that the design configurations developed during initial construction and revised during subsequent plant operations in Unit 1 were accurately maintained.
7.
Engineering Report ER 93-R-1002-01, "ANO 1 BWST Outlet Vortex Suppre aor,"
dated February 5,1993, was not properly verified for adequacy of the design in that it failed to consider the effects of instrument error in the analysis of the net positive suctici head for emergency core cooling system pumps taking suction on the borated water storage tank under vortexing conditions.
This is a Severity Level IV violation (Supplement 1) (50-313/9721-02).
Pursuant to the provisions of 10 CFR 2.201, Entergy Operations, Inc. is hereby required to submit a written statement or explaration to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20S55 with a copy to the Regional Administrator, Region IV,611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011, and a copy to the NRC Resident inspector at the facility that is the subject of this Notico, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a
" Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation or severity level, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to
3 avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response, if an adequate reply it not received within the time specified in this Notice, an ordel or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.
Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information, if you request withholding of such material, you.must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information), if safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Dated at Arlington, Texas this 6th day of February 1998